PURPOSES : The turning movement of vehicles is directly affected by such factors as vehicle length, wheelbase, steering angle, articulated angle, and wheel steering. Therefore, it is necessary to analyze the impact of changes in each factor on the turning of the vehicle. Because a vehicle with a long body, such as an articulated bus, makes a wide turn, this study analyzes the swept path of the driving vehicle considering the specifications of the vehicle. METHODS : This study was conducted by dividing driving routes into four routes of two-lane four-way roundabouts, and the turning conditions were examined for six types (Type 1–6) that simulated actual articulated bus data. The same vehicle specifications as those of the actual articulated bus were applied to the road design simulation (AutoTURN Pro), and the width of the swept path for the articulated bus was investigated based on the wheel steering control. Using a virtual reference line for dividing the inscribed circle into lanes of the roundabout by 5°, the driving width of the swept path was measured and the angle at which the driving width was largest during driving through the turning intersection was examined. In addition, the changes in the driving width of the swept path according to the wheel steering control under the same wheel turning conditions, as well as the articulated and steering angles, were investigated. RESULTS : The driving width of the swept path for the vehicle (Type 1) with the front wheel control function being an all-wheel system was less than that of an articulated bus with the largest driving width of 15° after entering the roundabout and 15° before entering the roundabout (Type 2). Furthermore, although the specifications of the vehicles were the same, it was determined that Type 5 was superior to Type 6 after reviewing the driving width in light of changes in the steering and articulated angles. CONCLUSIONS : The results of this study are expected to contribute to the field of road design considering traffic safety when large vehicles, such as articulated buses, turn on roundabouts or curved road sections.

낙동강 하구 기수생태 복원이 본격으로 논의가 진행 전인 2016년까지는 하류 수위의 예측을 위해 하구에서 수km 떨어진 기존 조위관측소(부산 및 가덕도)의 측정 자료를 활용하여 분석을 수행하였지만, 조위와 위상 차이로 인해 예측이 용이하지 않았다. 따라서, 낙 동강 하굿둑 인접 외해역에서 조석 영향을 받는 수위관측치를 이용하여 조석조화분해를 통한 정밀한 조위 예측 산정의 필요성이 대두되 어 본 연구를 수행하였다. 연구의 방법으로는 낙동강하굿둑 인근 외해역에서 10분 간격으로 기간별 관측자료의 저장상태 및 이상자료 유 무를 확인하고, 조석조화분해 프로그램인 TASK2000(Tidal Analysis Software Kit) Package를 이용하여 관측조위와 예측조위를 1대 1 비교하여 회귀상관분석을 수행하였다. 분석 결과, 관측조위와 예측조위간의 상관도는 0.9334로 높게 나타났으며, 당해 연도의 조위예측 분석시 직전 연도의 1년 조석관측 자료를 조화분해하여 산출된 조화상수를 이용하여 조위예측을 실시하면 보다 정확한 결과를 산출할 수 있음을 확인 하였다. 이를 바탕으로 2022년 예측조위를 생성하여 낙동강 하구 기수생태 복원의 해수유입량의 산정에 활용 중이다.

본 논문은 슬로싱 상태에 놓인 포화 상태 액체수소탱크에서 열 유속 및 BOG(Boil-off gas)의 경향을 다루고 있다. 특히, 액체-기체 간의 침투 및 혼합에 의한 열 교환에 관심을 두었다. 먼저, VOF(Volume of fluid)와 Eulerian 기반의 다상 유동모델로 모형 슬로싱 실험 을 모사하여 압력을 예측하고 계측된 값과 비교하였다. 자유 수면 및 충격 압력 실험 결과와 해석 결과를 비교하였으며, 유체의 속도 예측에서 정확할 수 있음을 간접적으로 증명하였다. 그리고 2차원의 Type-C 원통형 수소탱크를 대상으로 다상열유동해석을 수행하 였다. 이때 포화상태에 놓인 액체 및 기체수소를 가정하고, 해석을 통해 각 상간의 혼합에 의한 열 교환의 수준을 확인하고자 하였다. 단, 상간의 열 교환만을 관심으로 두고 있었으므로 질량전달 및 기화모델은 해석에서 제외하였다. 최종적으로 상의 혼합으로 인해 액 체수소로 유입되는 열 유속의 기여도에 대하여 정리하였다. 또한 액체수소로 유입되는 열 유속과 집중 질량 기반의 간이식을 통해 BOG 발생량 및 경향을 예측하고 분석하였다.

Attempts to use the molten salt system in various aspects such as MSR or energy storage systems are increasing. However, there are limitations in the molten salt-assisted technique due to the harsh corrosiveness of the molten salt, and a more detailed study on salt-induced corrosion is needed to solve this problem. In this study, corrosion behaviors of 80Ni-20Cr alloy in various salt environments such as eutectic NaCl-MgCl2 with NiCl2, CrCl2, and EuCl3 additives were investigated. Meanwhile, the corrosion acceleration effects of 80Ni-20Cr specimens were analyzed for various ceramic materials such as SiC, Al2O3, SiO2, graphite, and BN, and metallic materials such as Ni-based alloy, Fe-based alloy, and pure metals in a molten salt environment. The experiments were conducted at 973 K for up to 28 days, and after the experiment, the microstructural change of the specimen was analyzed through SEM-EDS, and salt condition was analyzed by ICP-OES.

The sorption/adsorption behavior of radionuclides, usually occurring at the solid-water interface, is considered to be one of the primary reactions that can hinder the migration of radiotoxic elements contained in the spent nuclear fuel. In general, various physicochemical properties such as surface area, cation exchange capacity, type of radionuclides, solid-to-liquid ratio, aqueous concentration, etc. are known to provide a significant influence on the sorption/adsorption characteristics of target radionuclides onto the mineral surfaces. Therefore, the distribution coefficient, Kd, inherently shows a conditiondependent behavior according to those highly complicated chemical reactions at the solid-water interfaces. Even though a comprehensive understanding of the sorption behavior of radionuclides is significantly required for reliable safety assessment modeling, the number of the chemical thermodynamic model that can precisely predict the sorption/adsorption behavior of radionuclides is very limited. The machine-learning based approaches such as random forest, artificial neural networks, etc. provide an alternative way to understand and estimate complicated chemical reactions under arbitrarily given conditions. In this respect, the objective of this study is to predict the sorption characteristics of various radionuclides onto major bentonite minerals, as backfill materials for the HLW repository, in terms of the distribution coefficient by using a machine-learning based computational approach. As a background dataset, the sorption database previously established by the JAEA was employed for random forest machine learning calculation. Moreover, the hyperparameters such as the number of decision trees, the number of variables to divide each node, and random seed numbers were controlled to assess the coefficient of determination, R2, and the final calculation result. The result obtained in this study indicates that the distribution coefficients of various radionuclides onto bentonite minerals can be reliably predicted by using the machine learning model and sorption database.

For deep geological repository of the spent nuclear fuel, the fuel assemblies loaded in the storage cask are transferred to the disposal cask and the operation is performed in the fuel handling hot cell at the fuel re-packaging facility. As the fuel handling hot cell shielding is accomplished by the concrete wall and the viewing glass window, the required shielding thickness was evaluated for both materials. The ordinary concrete is applied to hot cell wall and two kinds of glasses, i.e., single layer of lead glass and double layer of lead glass and borosilicate glass, are considered for the viewing glass window. A bare spent PWR fuel assembly exposed to the environment in the hot cell was considered as the neutron and gamma radiation sources. The neutron and gamma transport calculations were performed using the MAVRIC program of the SCALE code system for the dose rate evaluation. The dose limit of 10 μSv/h is applied as the target dose to establish the required shielding thickness. The concrete wall of 94 cm thickness reduces the total dose rate to 6.9 μSv/h, which is the sum of neutron dose and gamma dose. Penetrating the concrete wall, both of the neutron dose and the gamma dose decrease constantly with shield thickness and the gamma dose is always dominant through whole penetrating distance. Single layer lead glass of 74 cm thickness reduces total dose rate to 6.2 μSv/h. Applying double layer shield glass combined of lead glass and borosilicate glass, the total dose rate reduces to 3.6 μSv/h at same shield thickness of 74 cm. Through the shield glass, gamma dose decreases rapidly and neutron dose decreases slowly compared with those for concrete wall. In result, neuron dose becomes dominant on the window glass shielding. The more efficient dose reduction of double layer glass is achieved by the borosilicate glass’s superior neutron shielding power. Thus, the use of double layer glass of lead glass and borosilicate glass is recommended for the viewing glass of the fuel handling hot cell. Finally, it is concluded that about 1 m thick concrete wall and 75 cm thick viewing glass window are sufficient for the radiation shielding of the hot cell at the spent fuel repackaging facility.

The goal of the decommissioning of nuclear facilities is to remove the regulations from the Nuclear Safety Act. The media that can be considered at the time of remediation stage may usually include soils, buildings, and underground materials. In addition, underground materials may largely be the groundwater, buried pipes, and concrete structures. In fact, it can be seen that calculations of the Derived Concentration Guideline Level (DCGL) and ALARA action levels was conducted in the case of overseas decommissioning experiences of Nuclear Power Plants (NPPs). Therefore, the aim of this study is to review the remediation activities and scenarios applied for the calculation of ALARA action level from the overseas decommissioned nuclear power plants. Media that can be considered for DCGL calculation at the time of license termination may differ from site to site. If the DCGL for the target media was derived, whether additional remediation actions are required under the DCGL value from the ALARA perspective was identified by calculating the ALARA action levels in the case of the U.S. The activities to determine whether additional clean-up is justified under the regulatory criteria are remediation actions which is dependent on the material contaminated. Therefore, the typical materials that can be subjected to remediation are soils and structure basements in the overseas cases. Remediation actions involved in the decommissioning process on the structure surfaces can be typically considered to be scabbling, shaving, needle guns, chipping, sponge and abrasive blasting, pressure washing, washing and wiping, grit blasting, and removal of contaminated concrete. For the cost-benefit analysis of the media subject to DCGL calculation, it is necessary to assume a scenario for the remediation actions of the target media. The scenarios can be largely divided into two types. Those are basement fill and building occupancy scenario. In basement fill mode, buildings and structures on the site are removed, and the effect of receptors from the contamination of the remaining structures is considered. In the building occupancy mode, it is assumed that the standing building remains on the site after the remediation stage. It is a situation to evaluate how the effect of additional remediation actions changes as the receptors occupy inside of the contaminated building. Therefore, parameters such as population density, area being evaluated, monetary discount rate, numbers of years, etc. can be set and assessed according to the scenarios.

In KAERI, Waste storage facility in the radiation management area has stored a large amount of wood waste. The amount of waste is approximately 27,000 kg, it accounts for 17% of the total waste in waste storage facility. Proper disposal of wood waste improves the fire resistance performance, secure storage space and reduce disposal costs. In order to self-disposal of wood waste, it is necessary to satisfy the self-disposal standards stipulated by the domestic Atomic Energy Act and the treatment standards of the Waste Management Act. The main factors of standards are surface contaminant, radionuclide activity and radiation dose effects. To confirm the contamination of wood waste, direct indirect measurement methods and gamma nuclide analysis were performed. To evaluate radiation dose, various computational programs were used. The results of the analysis were satisfied with domestic regulations on the classification and self-disposal of radioactive wastes. Based on this results, KAERI submitted the report on wood waste self-disposal plan to obtain approval. After final approval, wood waste is to be incinerated and incineration ash is to be buried in the designated place. The objective of this study is to provide total procedure of wood waste self-disposal and effective representative sampling method.

During and after the construction of LILW disposal facilities, the decrease of groundwater head potential has been monitored. In addition, an increase of the electrical conductivity (EC) has been observed in several monitoring wells installed along the coastal coastline. Monitoring activity for groundwater head potential and hydrogeochemical properties is important to reduce the uncertainty in the evaluation of groundwater flow characteristics. However, the data observed in the monitoring wells are spatial point data, so there is a limit to the dimension. Several researchers evaluated groundwater head potential changes and seawater intrusion (SWI) potential for disposal sites using groundwater flow modeling. In case of groundwater flow modeling results for SWI, there is a spatial limit in directly comparing the EC observed in the monitoring wells with the modeling results. In a recent study, it was confirmed that the response of the long-range ground penetraiing radar (GPR) system was severely attenuated in the presence of saline groundwater. In order to reduce the spatial constraint of the groundwater monitoring wells for SWI, the characteristics of SWI within the disposal facility site by using the the results of a recent study of the long-range GPR system were investigated and evaluated in this study.

Engineered barriers (concrete and grout) in Low- and Intermediate-Level Waste (L/ILW) disposal facilities tend to degrade by groundwater or rainfall water over a long period of time. During the degradation process, radionuclides stored in the disposal facility might be released into the pore water, which can pass through the natural rock barriers (granite and sedimentary rock) and may reach the near-field and far-field. In this transportation, radionuclide might be sorbed onto the engineered and natural rock barriers. In addition, the organic complexing agent such as ethylenediaminetetraacetic acid (EDTA) and α-isosaccharinic acid (ISA), is also present in pore water, which may affect the sorption and mobility of radionuclide. In this study, the sorption and mobility of 90Sr under different conditions such as two pHs (7 and 13), different initial concentrations of organic complexing agents (from 10-5 M to 10-2 M), and solutions (groundwater, pore water, and rainfall water) were investigated in a batch system. The groundwater was collected at the L/ILW disposal facility located at Gyeongju in South Korea. The pore water and rainfall water were artificially made in the laboratory. The concrete, grout, granite, and sedimentary rock samples were collected from the same study sites from where the groundwater was collected. The rock samples were crushed to 53-150 micrometers and were characterized by XRD, XRF, SEM-EDS, BET, and zeta potential analyzer. 90Sr concentration was determined using liquid scintillation counting. The sorption of 90Sr was described by distribution coefficients (Kd) and sorption reduction factor (SRF). In the case of EDTA, the Kd values of 90Sr remained constant from 10-5 M to 10-3 M and tended to decrease at 10-2 M, while in case of ISA the Kd values decreased steadily as the concentration of ISA was increased from 10-5 M to 10-3 M; However, a sudden reduction in the Kd values were observed above 10-2 M. In comparison to EDTA, ISA gave a higher SRF of 90Sr. Therefore, from the above results, it can be concluded that the presence of ISA has a greater effect on the sorption and mobility of radionuclide in the solutions than EDTA, and the radionuclide may reach near- and far-field of the L/ILW disposal facility.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

Bentonite has been considered as a buffer material in a deep geological repository for high-level radioactive waste (HLW). Bentonite may come into contacted with various chemical solutions during the long-term storage. In particular, solutions containing K+ can affect stability of bentonite (e.g., illitization). The bentonite can be gradually saturated with the inflow of groundwater, and the temperature can rise simultaneously due to the decay of HLW. This study aimed to evaluate the bentonite stability in contacted with very highly concentrated K+ solutions with different pHs at 150°C. Batch reaction tests using KJ-II bentonite were performed for 30–150 days in teflon-stainless steel reactors. De-ionized (DI) water (pH = 6.0) and 1 M KCl (pH = 6.0), and 1 M KOH (pH = 12.5) solutions were used as reaction solutions. After completing batch reaction tests, the reacted samples were analyzed using various analytical techniques. For DI water, chemical, mineralogical, and physicochemical properties of reacted samples were similar to those of unreacted samples. For 1 M KCl solutions, cation exchage for Ca by K and slight changes in mineralogical properties of reacted samples were observed, but there are no significant changes in the physicochemical properties. In contrast, for 1 M KOH solutions, changes in chemical, mineralogical, and physicochemical properties of reacted samples were observed. Results of X-ray diffraction (XRD) analysis indicated dissolution of montmorillonite and formation of zeolite minerals, which were confirmed by thermogravimetricdifferential thermal analysis (TGA-DTA) and fourier transform infrared (FTIR) analysis. These results suggest that highly concentrated K+ (1 M) solution combined with high pH (12.5) and high temperate (150°C) may affect bentonite alteration. These prelimiary experiments were intended to qualitatively evaluate the mechanism and influncing factors of the buffer material alteration under extreme experimental conditions, and it is revealed that the conditions do not reflect the actual repository environment.

In high-level radioactive waste disposal, a high temperature is generated from the canister containing the waste in the engineered barrier, while groundwater flows into the buffer system from the host rock. The temperature increase and groundwater inflow result in the water phase change and saturation variation. Saturation change is related to the thermal conductivity of buffer material; hence the phase change and saturation strongly interact with the temperature evolution. The complex coupled behavior affects the stability of the whole disposal system, and the security of the repository is critical to human-being life. However, it is difficult to predict the long-term coupled behavior in the disposal system due to the considerable field-test scale, and therefore a numerical simulation is a suitable method having repeatability and cost-effectiveness. DECOVALEX is an international cooperating project for developing numerical methods and models for thermo-hydro-mechanical-chemical (THMC) interaction. DECOVALEX has a four-year cycle with various topics. At the current phase, Task C aims to simulate the full-scale emplacement (FE) experiment performed at Mont Terri underground rock laboratory. Nine research groups are participating in the task, and among them, KAERI simulates the experiment using OGS-FLAC. The simulator combines OpenGeoSys for TH simulation and FLAC3D for M simulation. Through the benchmark simulation, we verified OGS-FLAC for the two-phase flow analysis in the disposal system and finally modeled the FE experiment with a three-dimensional grid. We performed a simple sensitivity analysis to investigate the effect of input parameters on the two-phase flow system and confirmed that the compressibility and permeability affected the flow behavior. We also compared the simulation results to the field data and obtained well-matched results from a series of simulation.

The purpose of this study is to provide lessons learned in the experience of improvement work of fuel handling equipment at operating nuclear power plants. The upgrade of fuel handling equipment for safety enhancement and performance improvement has been going on for 15 years since the early 2000’s. The main goal is to increase fuel loading/unloading capability of the equipment from about 2.5 fuel assemblies per hour to more than six (6). It is achieved with sequential operations of three (3) fuel handling equipment, which consists of the refueling machine, the fuel transfer system and the spent fuel handling machine. The scope of the upgrade for fuel handling equipment is summarized as follows. The PC data control system based on PLC for interlocks and high speed motor drive system is introduced for better operating efficiency. The motors and drives for bridge, trolley, and hoist are replaced with AC servomotors and drivers, respectively. The fuel transfer system has an auto-initiation feature operating from the refueling machine or the spent fuel handling machine. The winch motor and drive for the carriage of fuel transfer system is also replaced with AC servomotors and drivers. And some of HPU (hydraulic power units) equipment for each building (reactor containment building and fuel handling building) are replaced to improve their function. The considerations for improvement of fuel handling equipment are as belows. 1) Fuel handling should be consistent with the instructions provided by the fuel designer and/or manufacturer, which are for Standard type fuel and Westinghouse type fuel, used in domestic nuclear power plants. Each fuel has unique design characteristics, which are PLC setpoints for overload and underload, slow speed zones for a bridge, trolley and hoist, allowable acceleration/deceleration value in handling, hoist elevation and manual speed in off-index operation at reactor. 2) The interlock system should be designed in accordance with design criteria specified by the utilities of nuclear power plant. 3) Performance should be improved according to the operating characteristics of the fuel handling equipment. High-speed and optimization of FTS upender and carriage are essential for operating performance so that its modification should be considered first. And the low speed and range in the operation mechanism of the hoist should be designed to comply with guidelines. 4) The accident analysis through self-diagnosis function and operation history in modification at domestic operating nuclear power plants would be good lessons learned. It is advisable to utilize such various information as it can help to improve reliability of nuclear fuel handling operation and power plant operation rate.

One of the promising candidates for heat transfer fluid is molten chloride salts. They have been studied in various fields such as the electrolyte of pyroprocessing, the molten salt reactor coolant, and the energy storage system media. Main considerations for utilizing molten chloride salts are the compatibility of salts with structural materials. The corrosion behavior of structural materials in molten chloride salts must be understood to identify suitable materials against the corrosive environment. In this study, the corrosion behavior of a candidate structural material, Hastelloy N, in molten LiCl- KCl salt at 500°C were investigated by the electrochemical impedance spectroscopy (EIS) method. The sheet type of Hastelloy N was utilized as the working electrode in LiCl-KCl to measure the EIS data for 100 hours with 5 hours of time intervals. The EIS data were measured in the frequency range from 104 Hz to 10-2 Hz with the AC signal (amplitude = 20 mV) at open circuit potential. The capacitance semicircle observed in Nyquist plots for all periods indicates that charge-transfer controlled reactions occur. As the immersion time progresses, the radius of the semicircle in Nyquist plots and the impedance and phase angle in Bode plots decrease. These behaviors suggest a decreasing reaction resistance and the corrosion reactions are accelerated with the immersion time. The EIS data were fitted using the equivalent circuit to achieve quantitative results. Two capacitor-resistor components were considered due to the overlapped shape of two valleys in phase angle. The depressed shape of the semicircle in Nyquist plots led to the use of the constant phase element(Q) instead of the capacitor. Therefore, R(Q(R(QR))) circuit was selected to fit the EIS data. Fitting results show that the charge transfer resistance decreases dramatically within 1 day and then converges. The film resistance shows no clear trends, but the increase of the film admittance value indicates the decreased film thickness. Consequently, the film appears to exist like the oxide layer but it does not act as a protective layer. The real-time EIS data were measured in molten salt and provides the corrosion behavior over time. The corrosion mitigation strategy should consider that the corrosion of Hastelloy N accelerates over time and its intrinsic film cannot act as the protective layer. The next steps of this study are to evaluate other candidate structural materials and to demonstrate the presence of the film.

PURPOSES : Aiming to evaluate the consistency of road markings at night, the luminance values of road markings at night were compared for merging and diverging areas. METHODS : To evaluate the consistency of road marking at nighttime in merging and diverging areas, the highway geometric structures and luminance of the road markings were collected and analyzed together at the merging and diverging areas. "Luminance" in this study referred to the overall luminance of the road markings as reflected together from surrounding lights, such as moonlight or artificial lights. The luminance of the road markings 90 m ahead of the driver were used. The measured luminance values were analyzed based on the difference ratios and an analysis of variance. RESULTS : Based on a grouping of three categories (interchanges (ICs), merging, and diverging areas), it was found that the difference ratios and analysis of variance values from the ICs and merging and diverging areas were not consistently acceptable. CONCLUSIONS : After evaluating the consistency of road markings at night in the merging and diverging areas, it can be concluded that there is a need for more consistency in the luminance on merging and diverging areas on highways. To enhance consistency, more dedicated lighting guidance for merging and diverging areas on highway areas may be necessary.